ML041620247
| ML041620247 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 06/11/2004 |
| From: | Gregory Suber NRC/NRR/DRIP/RLEP |
| To: | Forbes J Entergy Operations |
| Suber G, NRR/DRIP/RLEP 301-415-1124 | |
| References | |
| TAC MB8402 | |
| Download: ML041620247 (13) | |
Text
June 11, 2004 Mr. Jeff Forbes Vice President, Operations ANO Entergy Operations, Inc.
1448 S. R. 333 Russellville, AR 72801
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE ARKANSAS NUCLEAR ONE, UNIT 2, LICENSE RENEWAL APPLICATION (TAC NO. MB8402)
Dear Mr. Forbes:
The U.S. Nuclear Regulatory Commission (NRC) is reviewing a license renewal application (LRA) submitted by Entergy Operators Inc. (Entergy or the applicant) dated October 14, 2003 for the renewal of the operating licenses for Arkansas Nuclear One, Unit 2, pursuant to Title 10 Code of Federal Regulations Part 54 (10 CFR Part 54). The NRC staff has identified, in the enclosure, areas where additional information is needed to complete the review.
Specifically, the enclosed request for additional information (RAI) is from Section 3.1 Reactor Vessel, Internals and Reactor Coolant System, Section 4.2 Reactor Vessel Neutron Embrittlement, Section 4.7 Other Plant-Specific Time-Limited Aging Analysis, and Appendix B, Sections B.1.1 Alloy 600 Aging Management, B.1.3 Boric Acid Corrosion Prevention, B.1.21 Reactor Vessel Integrity, B.1.22 Reactor Vessel Internals Cast Austenitic Stainless Steel Components, and B.1.23 Reactor Vessel Internals Stainless Steel Plates, Forgings, Welds, and Bolting. These RAIs have been discussed with your staff.
Your response to these RAIs is requested within 45 days from the date of this letter. If you have any questions, please contact me at (301) 415-1124 or e-mail gxs@nrc.gov.
Sincerely,
/RA/
Gregory F. Suber, Project Manager License Renewal Section A License Renewal and Environmental Impacts Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No.: 50-368
Enclosure:
As stated cc w/encl: See next page
June 11, 2004 Mr. Jeff Forbes Vice President, Operations ANO Entergy Operations, Inc.
1448 S. R. 333 Russellville, AR 72801
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE ARKANSAS NUCLEAR ONE, UNIT 2, LICENSE RENEWAL APPLICATION (TAC NO. MB8402)
Dear Mr. Forbes:
The U.S. Nuclear Regulatory Commission (NRC) is reviewing a license renewal application (LRA) submitted by Entergy Operators Inc. (Entergy or the applicant) dated October 14, 2003 for the renewal of the operating licenses for Arkansas Nuclear One, Unit 2, pursuant to Title 10 Code of Federal Regulations Part 54 (10 CFR Part 54). The NRC staff has identified, in the enclosure, areas where additional information is needed to complete the review.
Specifically, the enclosed request for additional information (RAI) is from Section 3.1 Reactor Vessel, Internals and Reactor Coolant System, Section 4.2 Reactor Vessel Neutron Embrittlement, Section 4.7 Other Plant-Specific Time-Limited Aging Analysis, and Appendix B, Sections B.1.1 Alloy 600 Aging Management, B.1.3 Boric Acid Corrosion Prevention, B.1.21 Reactor Vessel Integrity, B.1.22 Reactor Vessel Internals Cast Austenitic Stainless Steel Components, and B.1.23 Reactor Vessel Internals Stainless Steel Plates, Forgings, Welds, and Bolting. These RAIs have been discussed with your staff.
Your response to these RAIs is requested within 45 days from the date of this letter. If you have any questions, please contact me at (301) 415-1124 or e-mail gxs@nrc.gov.
Sincerely,
/RA/
Gregory F. Suber, Project Manager License Renewal Section A License Renewal and Environmental Impacts Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No.: 50-368
Enclosure:
As stated cc w/encl: See next page DISTRIBUTION: See next page Accession No: ML041620247 Document Name: C:\\ORPCheckout\\FileNET\\ML041620247.wpd OFFICE:
LA:RLEP PM:RLEP SC:RLEP NAME:
M. Jenkins G. Suber S. Lee DATE:
5/21/04 6/4/04 6/11/04 OFFICIAL RECORD COPY
Enclosure ANO-2 LICENSE RENEWAL APPLICATION FOR REACTOR VESSEL, INTERNALS AND REACTOR COOLANT SYSTEM REQUEST FOR ADDITIONAL INFORMATION (TAC NO. MB8402)
Table 3.1.2-1 Reactor Vessel and CEDM Pressure Boundary RAI 3.1.2-1.1 The staff requests additional information on the applicants aging management reviews (AMRs) for managing cracking in low alloy steel components that are exposed to an external air environment. Aging management strategies for license renewal are somewhat dependent on the specific types of aging mechanisms that can induce age-related degradation, and not specifically on the general classification of the aging effect. For the low alloy steel components in the reactor coolant system, confirm that cracking is an applicable aging effect requiring aging management. Specifically, define which aging mechanism or mechanisms are known to induce cracking in low alloy steel components that are exposed to an external air environment.
RAI 3.1.2-1.2 The staff requests additional information on the applicants AMRs for managing loss of material in nickel-based alloy components that are exposed to an internal environment of treated borated water. For the nickel-based alloy components in the reactor coolant system, define which aging mechanism or mechanisms are known to induce loss of material in nickel based alloy components that are exposed to an internal environment of treated borated water and explain how this aging effect will be addressed by the aging management strategy for license renewal.
RAI 3.1.2-1.3 The staff requests additional information on the applicants AMRs for managing loss of material in stainless steel components that are exposed to an internal environment of treated borated water. For the stainless steel components in the reactor coolant system, define which aging mechanism or mechanisms are known to induce loss of material in stainless steel components that are exposed to an internal environment of treated borated water and explain how this aging effect will be addressed by the aging management strategy for license renewal.
RAI 3.1.2-1.4 Clarify where the boric acid corrosion aging mechanism is considered in Section 3.1 of the License Renewal Application (LRA), and in Tables 3.1.2-1 and 3.1.2-3. Specify which component types, materials, environments, aging effects requiring management, and aging management programs are associated with this aging mechanism and explain how this aging effect will be addressed by the aging management strategy for license renewal.
Table 3.1.2-2 Reactor Vessel Internals RAI 3.1.2-2.1 The staff requests additional information on the applicants AMRs for managing loss of material and cracking in cast austenitic stainless steel (CASS) components that are exposed to an internal environment of treated borated water. For the CASS components in the reactor coolant system, define which aging mechanism or mechanisms are known to induce loss of material and cracking in CASS components that are exposed to an internal environment of treated borated water and explain how this aging effect will be addressed by the aging management strategy for license renewal.
Table 3.1.2-3 Class 1 Piping, Valves, and Reactor Coolant Pumps RAI 3.1.2-3.1 In Table 3.1.2-3, on page 3,1-79, the applicant identifies treated water as the external environment for the reactor coolant pump thermal barrier heat exchanger inner coil. In addition, on page 3,1-80, the applicant identifies treated water as the internal environment for the reactor coolant pump thermal barrier heat exchanger outer coil and bored hole heat exchanger. Loss of material, cracking, and fatigue are defined as the aging effects requiring management.
The aging management programs (AMPs) identified to manage these aging effects are Inservice Inspection and Time-Limited Aging Analysis (TLAA)-Metal Fatigue. The applicant's Auxiliary Systems Water Chemistry Control AMP, described in Section B.1.30.1, identifies its purpose as managing loss of material, cracking, and fouling of components exposed to treated water systems. The applicant has identified similar components of the same material which are exposed to the same environment as being managed by a Water Chemistry AMP and referenced concurrence with NUREG-1801, VII.C2.2-a. Provide justification for excluding an AMP to manage the water chemistry of the treated water environment as applicable to these components.
Table 3.1.2-4 Pressurizer RAI 3.1.2-4.1 The staff requests additional information on the applicants AMRs for managing cracking in carbon steel components that are exposed to an external air environment (i.e. support skirt).
For the carbon steel components in the reactor coolant system, define which aging mechanism or mechanisms are known to induce cracking in carbon steel components that are exposed to an external air environment and explain how this aging effect will be addressed by the aging management strategy for license renewal.
RAI 3.1.2-4.2 The staff requests additional information on the applicants AMRs for managing cracking in stainless steel components that are exposed to an external air environment (i.e. mechanical nozzle seal assembly clamp bolting [studs, nuts, washers]). For the stainless steel components in the reactor coolant system, define which aging mechanism or mechanisms are known to induce cracking in stainless steel components that are exposed to an external air environment and explain how this aging effect will be addressed by the aging management strategy for license renewal.
RAI 3.1.2-4.3 The staff requests additional information on the applicants AMRs for managing loss of material and cracking in low alloy steel clad with stainless steel and nickel based alloy components that are exposed to an internal environment of treated borated water. For the low alloy steel clad with stainless steel and nickel based alloy components in the reactor coolant system, define which aging mechanism or mechanisms are known to induce loss of material and cracking in low alloy steel clad with stainless steel and nickel based alloy components that are exposed to an internal environment of treated borated water and explain how this aging effect will be addressed by the aging management strategy for license renewal.
RAI 3.1.2-4.4 Table 3.1.2-4, Page 3,1-84 identifies the pressurizer lower head, lower shell, upper shell, and upper head as component types. The applicant identified the aging effect of loss of material, and specified that it is applicable to the unclad low alloy steel of the lower head only. Provide justification for limiting the aging effect to only the lower head since many components of the pressurizer are susceptible to boric acid corrosion in a treated borated water environment, and would require that the aging effect of loss of material is managed.
RAI 3.1.2-4.5 Recent operational experience at both domestic and foreign facilities (Palo Verde Units 2 and 3, Millstone Unit 2, Waterford Unit 3, and Tsuruga Unit 2 in Japan) has shown that leakage of pressurizer penetrations due to primary water stress corrosion cracking (PWSCC) is an aging effect that requires management. Since AMP B.1.19 Pressurizer Examinations is limited only to managing cracking of the stainless steel and nickel-based alloy cladding and attachment welds by examination of the adjacent base metal, discuss how the aging effect of PWSCC will be managed for the pressurizer penetrations for the period of extended operation at ANO-2.
Include scope, frequency, technique, acceptance criteria, and the technical basis for future examinations.
Section 4.2 Reactor Vessel Neutron Embrittlement RAI-4.2-1 The applicant assumes a capacity factor of 80% for the time limiting aging analyses associated with reactor vessel neutron embrittlement that are described in Section 4.2 of the LRA. These evaluations are based on end-of-license (EOL) fluences corresponding to 48 effective power years (EFPY). Staff reviews of current and future trends for plant operations in the nuclear power industry indicate capacity factors of 90% or greater for many plants. Provide justification for the estimated 48 EFPY fluence for ANO-2. If the estimated 48 EFPY fluence cannot be justified, provide results of revised evaluations for reactor vessel neutron embrittlement at higher levels of fluence projected to the end of the period of extended operation.
RAI 4.2-2 Pursuant to 10 CFR Part 54.21(d), the FSAR Supplement for a facility license renewal application (LRA) must contain a summary description for each aging management program and time-limited aging analysis (TLAA) proposed for management of the effects of aging. The staff has determined that Appendix A of the LRA (FSAR Supplement) did not include a corresponding FSAR Supplement summary description for Table 4.2-2 in TLAA 4.2, Reactor Vessel Neutron Embrittlement of the LRA. Table 4.2-2 contains an evaluation of reactor vessel extended life for pressurized thermal shock. The staff notes that the corresponding table for the upper-shelf energy extended life evaluation (Table 4.2-1) was included in the FSAR Supplement. Pursuant to 10 CFR 54.21(d), the staff requests that a corresponding FSAR Supplement summary description for LRA Table 4.2-2 be included in the FSAR Supplement.
Section 4.7.1 Reactor Coolant System Piping Leak-Before-Break RAI-4.7.1-1 In Section 4.7.1 of the LRA, the applicant addresses the RCS piping LBB TLAA and concludes that the LBB evaluation for fatigue crack growth remains valid for the period of extended plant operation. How much additional crack growth was predicted by the updated calculations for the end of 60 years compared to that originally predicted for 40 years? What was the criteria or basis for concluding that this amount of additional crack growth was insufficient to exclude fatigue as a damage mechanism that would limit the application of LBB to ANO-2 RCS piping in accordance with the NRC guidance for LBB?
Section 4.7.2 Reactor Coolant Pump Code Case N-481 RAI 4.7.2-1 In Section 4.7.2 of the LRA, the applicant addresses the reactor coolant pump (RCP) Code Case N-481 time limiting aging analysis. The applicant used fully aged (saturated) properties in the analysis, and concluded that effects of thermal aging on material properties of cast austenitic stainless steel are addressed for the period of extended operation. Discuss whether these properties are the same bounding properties that were used for embrittled cast stainless materials assumed in the Combustion Engineering report, CEN-367-A which is an analysis for LBB. If other material properties were used, provide justification for the properties that were used for the Code Case N-481 analysis.
RAI-4.7.2-2 In Section 4.7.2 of the LRA, the applicant addresses the RCP Code Case N-481 time limiting aging analysis and concludes that the evaluation for fatigue crack growth remains valid for the period of extended plant operation. Discuss the additional crack growth that was predicted by the updated calculations at the end of 60 years, and compare the crack growth to that originally predicted for 40 years. Provide the criteria or basis for concluding that this amount of additional crack growth is sufficiently small to justify continued application of Code Case N-481.
Section 4.7.3 Reactor Coolant Pump Flywheel RAI 4.7.3-1 In 4.7.3 (Page 4.7-2) of the LRA, the applicant concluded that the RCP flywheel is not a TLAA.
The basis for this conclusion is a 1997 safety evaluation of a fatigue crack growth analysis that was presented in a Combustion Engineering Owners Group topical report. This safety evaluation allowed the licensee to lengthen the RCP flywheel inspection period for ANO Units 1 and 2 and five other units. The fatigue crack growth analysis for ANO Units 1 and 2 is based on 4,000 RCP startup and shutdown cycles. The RCP flywheel was identified as a TLAA in the LRA for ANO Unit 1, and two other units that are identified in the topical report and that have been granted renewed licenses.
Please provide justification why the RCP flywheel is not a TLAA for ANO-2. If the RCP flywheel is a TLAA, provide the TLAA for the RCP Flywheel for ANO Unit 2, and include the justification for why 4,000 RCP startup and shutdown cycles remain bounding through the end of the extended period of operation for ANO-2. In addition, the applicant must include an FSAR Supplement summary description, in Appendix A, of the LRA for the TLAA on fatigue-induced crack growth of the ANO-2 RCP flywheel. The summary description should include a discussion on the safety margin for the acceptable flaw size, and the justification for why 4,000 RCP startup and shutdown cycles remain bounding through the end of the extended period of operation for ANO.
Section 4.7.5 Alloy 600 Nozzle Repairs RAI-4.7.5-1 Demonstrate that the designs of repaired nozzles will have sufficient structural integrity against loss of material by corrosion and will meet their minimum wall thickness requirements through the expiration of the extended period of operation.
RAI 4.7.5-2 Justify and validate the CEOGs conclusion that growth of the existing flaw in the original Alloy 600 J-groove weld material by stress corrosion cracking is not a plausible effect during the period of extended operation.
B.1.1 Alloy 600 Aging Management RAI B.1.1-1 Confirm that all of the components listed in the Alloy 600 Aging Management Program are covered under the inservice inspection requirements of Section XI of the ASME Code, and for any components not covered by Section XI inservice inspection requirements, provide a complete description of the proposed inspections including a technical justification for the inspection method and frequency.
RAI B.1.1-2 The applicant stated that no preventative actions will be taken as part of the Alloy 600 Aging Management Program to prevent aging effects or mitigate aging degradation. The NRC staff notes that several preventive actions and common industry practices have been used to mitigate PWSCC associated with Alloy 600. Examples of these include: nickel plating of the surfaces of Alloy 600 components that are exposed to treated water, replacement of leaking Alloy 600 instrument nozzles with Alloy 690 material, preventive replacement of selected pressurizer and RCS piping instrument nozzles with Alloy 690 material, monitoring the electrochemical potential, and water chemistry control. Provide a description of any preventive actions and/or water chemistry monitoring programs ANO-2 is currently implementing that may be used to address the Alloy 600 cracking issue.
RAI B.1.1-3 In the Alloy 600 Aging Management Program under the program attribute, Detection of Aging Effects, the applicant states that the measurement, vent, upper level, and temperature nozzles, and heater sheath, heater sleeve, and end plug received visual examination (VT-2) from the exterior of the vessel in accordance with ASME Section XI, Examination Category B-P. For many of these components, the Alloy 600 pressure boundary welds are covered by insulation.
Service experience has shown that, early indications of through-wall leakage (e.g., boric acid on the component surface) are very difficult to detect when VT-2 examinations are performed with the insulation in place. Provide justification for not removing insulation when performing VT-2 examinations on the components mentioned above. In addition, provide the frequency of inspection, and the results of any volumetric non-destructive examination that has been performed.
RAI B.1.1-4 A.
In the Alloy 600 Aging Management Program under the program attribute, Detection of Aging Effects, the applicant states that guidance from the Materials Reliability Project (MRP) in conjunction with the PWR owners groups will be used to identify critical locations for inspection and augmentation of existing ISI inspections at ANO-2 where appropriate. The staff believes that the strategic plan developed by the industry will be comprehensive and recommendations may be applicable to all 10 elements of the Alloy 600 Aging Management Program. Identify the date that ANO-2 commits to submit, for review and approval, an augmented aging management program that includes all recommendations from the industrys strategic plan, and meets the 10 elements in accordance with the guidance in NUREG-1800, Appendix A.1, Aging Management Review - Generic, Table A.1-1, Elements of an Aging Management Program for License Renewal. The date must be prior to the period of extended operation.
B.
The Updated Final Safety Analysis Report (UFSAR) for ANO-2 does not contain a commitment to use guidance developed by the Electric Power Research Institute (EPRI)
MRP program, and to submit the inspection plan for review and approval. Confirm that the UFSAR will be revised to reflect the above mentioned commitment for management of Alloy 600 components.
RAI B.1.1-5 In the Alloy 600 Aging Management Program under the program attribute, Operating Experience, the applicant states that the Alloy 600 aging management program is a new program for which there is no specific operating experience for ANO-2. The staff is aware of several reported failures related to Alloy 600 welded components in other PWRs including several failures in other Combustion Engineering (CE) NSSS design. Specifically, PWSCC has been reported in Alloy 82/182 J-groove welds that are used to join Alloy 600 small bore nozzles to CE-designed pressurizers, steam generators, and/or hot legs. The staff believes it important for the applicant to review relevant industry service experience and incorporate lessons learned into the Alloy 600 program. Therefore, the applicant should discuss what industry initiatives it plans to follow in order to incorporate experience related to Alloy 600 into the ANO-2 Alloy 600 AMP.
B.1.3 Boric Acid Corrosion Prevention RAI B.1.3-1 Provide the basis for the proposed acceptance criteria that will be developed as part of the following enhancement to the Boric Acid Corrosion Prevention AMP:
The program acceptance criteria will be revised to address electrical components in addition to ferritic steel.
RAI B.1.3-2 In the Operating Experience Section of B.1.3, Boric Acid Corrosion Prevention, the applicant states that recent industry events regarding reactor vessel head degradation required assessments at each site to ensure that boric acid corrosion prevention programs are adequate and functioning effectively. The applicant also states that a self assessment was performed in February 2003, and no significant findings were identified during this assessment. Discuss how program revisions have incorporated lessons learned from the Davis Besse vessel head degradation and the control rod drive mechanism penetration cracking discussed in NRC Bulletins 2001-01, 2002-01, 2002-02, and NRC Order EA-03-009. Also, provide a discussion on implementation of corrective actions in the program to prevent reoccurrence of degradation caused by boric acid leakage, as required by Generic Letter 88-05.
B.1.21 Reactor Vessel Head Penetration RAI B.1.20-1 The applicant states that the Corrective Action Program was used to incorporate industry operating experience into the Reactor Vessel Head Penetration program, and to develop inspection requirements that are specific to ANO-2. The applicant also states that recent reactor vessel head penetration nozzle inspections were performed in accordance with the commitments in the ANO-2 response to NRC Bulletin 2001-01. The NRC staff notes that in February 2003, NRC Order EA-03-009 was issued. This order supercedes NRC Bulletins 2001-01 and 2002-01, and requires that licensees assess the susceptibility of the reactor vessel head to PWSCC-related degradation. The Order also requires the licensee to commit to an augmented inspection program for the reactor pressure vessel head based upon the susceptibility to PWSCC. Provide a commitment for implementation of the ANO-2 augmented inspection plan in accordance with NRC Order EA-03-009 for the period of extended operation and confirm that the UFSAR will be revised to reflect the above mentioned commitment for inspection of the reactor pressure vessel head.
B.1.21 Reactor Vessel Integrity RAI B.1.21-1 The description of this AMP does not include a specific reactor vessel specimen capsule withdrawal schedule for the period of extended operation. Please provide a specific schedule through the end of the period of extended operation for staff review. In addition, please revise the UFSAR Table 5.2-12 accordingly.
B.1.22 Reactor Vessel Internals Cast Austenitic Stainless Steel Components RAI B.1.22-1 The Reactor Vessel Internals Cast Austenitic Stainless Steel Components (CASS) AMP is currently not in place. The applicant states in LRA Section B.1.22, that the AMP will be consistent with NUREG-1801 (Generic Aging Lessons Learned [GALL]), and that it will initiate the program prior to the period of extended operation. Management of the aging effects associated with void swelling of PWR reactor vessel internals is not included in the GALL report. The staff requests that the applicant formally make a commitment to participate in industry initiatives, and to implement industry recommendations regarding void swelling when they become available. The staff also requests that the applicant submit the inspection program to manage the aging effects associated with void swelling to the NRC for review and approval no later than three years prior to the period of extended operation.
RAI B.1.22-2 The staff found some differences between Table 3.1-2 in NUREG-1800 (Standard Review Plan for License Renewal) and the ANO-2 LRA UFSAR Section A.2.1.23, Reactor Vessel Internals CASS Program. Table 3.1-2 in the Standard Review Plan provides a description of what should be included in the UFSAR Supplement for aging management of RV internals, and the reactor coolant system for license renewal reviews. The UFSAR should state that the inservice inspection program will be augmented to include enhanced examinations of non-bolted components, and other demonstrated acceptable methods for bolted components for certain susceptible or limiting components or locations. Clarify why the enhanced examination and/or component-specific flaw evaluation for the CASS component, which are specified in NUREG-1800, are not included in ANO-2 LRA UFSAR Section A.2.1.23.
B.1.23 Reactor Vessel Internals Stainless Steel Plates, Forgings, Welds, and Bolting RAI B.1.23-1 This Reactor Vessel Internals Stainless Steel Plates, Forgings, Welds, and Bolting AMP is currently not in place. The applicant states in LRA Section B.1.23, that the AMP will be consistent with NUREG-1801 (GALL), and that it will initiate the program prior to the period of extended operation. Management of the aging effects associated with void swelling of PWR reactor vessel internals is not included in the GALL report. The staff requests that the applicant formally make a commitment to participate in industry initiatives, and to implement industry recommendations regarding void swelling when they become available. The staff also requests that the applicant submit the inspection program to manage the aging effects associated with void swelling to the NRC for review and approval no later than three years prior to the period of extended operation.
RAI B.1.23-2 The staff noted that the water chemistry system and the enhanced examination of non-bolted components are not discussed in LRA UFSAR Supplement, Section A.2.1.24 as is stated in NUREG-1800, Table 3.1-2, page 3.1-27. Clarify why the water chemistry system and the enhanced examination of non-bolted components, which are specified in NUREG-1800, are not included in ANO-2 LRA UFSAR Section A.2.1.24.
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Arkansas Nuclear One, Unit 2 cc:
Executive Vice President
& Chief Operating Officer Entergy Operations, Inc.
P. O. Box 31995 Jackson, MS 39286-1995 Director, Division of Radiation Control and Emergency Management Arkansas Department of Health 4815 West Markham Street, Slot 30 Little Rock, AR 72205-3867 Winston & Strawn 1400 L Street, N.W.
Washington, DC 20005-3502 Mr. Mike Schoppman Framatome ANP, Richland, Inc.
Suite 705 1911 North Fort Myer Drive Rosslyn, VA 22209 Senior Resident Inspector U.S. Nuclear Regulatory Commission P. O. Box 310 London, AR 72847 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 County Judge of Pope County Pope County Courthouse Russellville, AR 72801 Vice President, Operations Support Entergy Operations, Inc.
P. O. Box 31995 Jackson, MS 39286-1995 Wise, Carter, Child & Caraway P. O. Box 651 Jackson, MS 39205 Garry Young 1448 SR 333 Russellville, AR 72802 Mr. Fred Emerson Nuclear Energy Institute 1776 I St., N.W., Suite 400 Washington, DC 20006-3708